Development of a computer code for neutronic modeling of reactor core with rectangular fuel assembly
Publish place: International Conference on New Research Findings in Electrical Engineering and Computer Science
Publish Year: 1394
نوع سند: مقاله کنفرانسی
زبان: English
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شناسه ملی سند علمی:
COMCONF01_255
تاریخ نمایه سازی: 8 آذر 1394
Abstract:
In this study, we have reformulated the AFEN method for diffusion equations and implemented in a version of the AFEN method code MGHANSP3 for rectangular-z geometry. For increasing the speed of calculations, the coarse group rebalancing (CGR) method has been utilized. Finally, we developed a computer code, named SDANM, using C# programming language. This code takes few group macroscopic cross sections and calculates the effective multiplication factor, reactivity, multi-groups fluxes, power density distributions and power peaking factor in reactor cores with square geometry.The solution accuracy is examined for IAEA benchmark problem. The numerical results illustrate that the SDANM code is an accurate code for calculating multiplication factor and power density distributions
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Authors
Mohammad Hassan Jalili Bahabadi
Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran
Ali Pazirandeh
Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran
Mitra Athari
Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran
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